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Abstract  

Indigenously synthesized extractant, phenyl (octyl) phosphonic acid (POPA) in tri-n-butylphosphate (TBP) and dodecane, has been investigated for the separation of americium from trivalent lanthanides in nitric acid medium as well as diethylene triaminepentaacetic acid (DTPA) and lactic acid mixture (TALSPEAK medium). Various experimental parameters like concentration of DTPA, lactic acid, TBP, nitrate ions and pH of the aqueous feed solution have been optimized to obtain the highest separation factor between americium and europium. Bulk actinide–lanthanide separation reagent, tetra (ethylhexyl) diglycolamide (TEHDGA), was equilibrated with simulated solution of americium and lanthanides, equivalent in concentration to the reprocessing waste originating from PHWR spent fuel. DTPA/lactic acid mixture was used to strip the metal ions from the loaded organic phase and re-extracted into POPA in TBP/dodecane to evaluate the separation factor of individual lanthanides with respect to americium. Very good separation factors between americium and trivalent lanthanides were obtained.

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Abstract  

Ion exchange studies of uranium(VI), thorium(IV), plutonium(IV) and europium(III) ions on a macroreticular cation exchange resin, Amberlyst A-15, from solutions of 30% and 5% TBP—Shell Sol-T have been carried out. The metal ions were extracted into TBP Shell Sol-T phase from 8M NH4NO3 at different nitric acid concentrations. Ion exchange distribution ratios as a function of organic phase acidity of 30% and 5% TBP have been computed. Separation factors computed from the observed Kd values are plotted as a function of organic phase acidity.

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Abstract  

Distribution ratios of europium(III), thorium(IV), uranium(VI) and plutonium(IV) ions on Amberlyst A-15, a macroreticular polystyrene sulfonate resin, after extraction in HTTA-TBP-Shell Sol-T and HTTA-TOPO-benzene solutions have been determined as a function of the aqueous acidity. The affinity orders were EuPu>Th>U and Eu>Th>Pu>U in the former and the latter solutions, respectively. Separation factors were computed from the observed Kd values. A procedure for the separation of a mixture of Eu(III), Th(IV), and U(VI) ions in HTTA-TOPO-benzene solution in an ion-exchange column is described.

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Abstract  

Batch equilibrium distribution data for Ce3+/H+ and Am3+/H+ systems in 0.2 to 4.0M HNO3 on 1, 2, 4, 8, 10, 12 and 16% crosslinked Dowex 50W resins are reported. These data, along with the mean ionic activity coefficients of the tracer in the mixed electrolyte solutions, were used to calculate the equilibrium constant (K a) uncorrected for the resin phase activity coefficient. The logK a values obtained at various ionic strengths were fit to a second order quadratic equation. Using the fitting parameters, logK a values were calculated for the different resins at zero ionic strength. LogK (equilibrium constant) values were computed by neglecting the changes in the activity coefficient terms in the resin phase due to resin loading. The logK values reported for Ce3+/H+ systems at a few crosslinkages are compared and the magnitude of the error in the approximate calculations is discussed.

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Abstract  

Back-extraction of tri- and tetravalent actinides from diisodecylphosphoric acid (DIDPA) is studied using hydrazine carbonate as back-extractant. In experiments using 0.5M DIDPA–0.1M TBP n-dodecane solution, Am(III), Eu(III), Pu(IV) and Np(IV) are back-extracted, and the distribution ratios are decreased with an increase of hydrazine carbonate concentration. The back-extraction equilibria are confirmed by slope analysis in consideration of neutralization between DIDPA and hydrazine carbonate, which occurs quantitatively during back-extraction. In particular, oxidation of Np(IV) to Np(V) during back-extraction is observed by measuring absorption spectra. The hydrazinium ion acts as an oxidation reagent in the back-extraction of Np(IV). Separation factors of those metals are compared with the results of HDEHP. Hydrazine carbonate back-extracts Np(IV) more selectively from DIDPA than from HDEHP.

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Abstract  

A generator system has been developed for the preparation of carrier-free 90Y from 90Sr present in the high level waste (HLW) of the Purex process by employing a supported liquid membrane (SLM) using 2-ethylhexyl-2-ethylhexyl phosphonic acid (KSM-17 equivalent to PC 88A) supported on a polytetrafluoro ethylene (PTFE) membrane. When uranium depleted Purex HLW at appropriate acidity is passed sequentially through octyl (phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) sorbed on chromosorb-102 (abbreviated as CAC) and Zeolite AR1 (synthetic mordenite) columns, all the trivalent, tetravalent and hexavalent metal ions and monovalent 137Cs ions are sorbed. After adjusting to pH 2 with NaOH the resulting effluent is used as feed in a single stage membrane cell partitioned with a PTFE membrane impregnated with KSM-17 and having a feed and receiver compartment with 5.0 ml capacity each. The receiver compartment was filled with a 0.5M HNO3 or 0.5M HCl stripping solution. 90Y alone is preferentially transported across the membrane leaving behind all the impurities viz. 90Sr, 125Sb, 106Ru, 106Rh, etc. in the feed compartment. This technique can yield 90Y in mCi levels in a pure and carrier-free form for medical applications. The feed can be reused repeatedly after allowing for 90Y buildup.

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Journal of Radioanalytical and Nuclear Chemistry
Authors: S. Ravi, S. Ravi, A. K. Deepa, A. K. Deepa, S. Susheela, S. Susheela, P. V. Achuthan, P. V. Achuthan, S. Anil Kumar, and U. Jambunathan

Summary  

A method has been developed for the estimation of 90Sr in reprocessed uranium oxide samples obtained from the Purex processing of natural uranium spent fuel discharged from the research reactor. The method employs a combination of precipitation and solvent extraction procedure to eliminate other beta-impurities prior to resorting to the estimation of 90Sr by beta-counting. 106Ru was eliminated by volatalizing with perchloric acid, uranium was removed by carrier precipitation with strontium as sulphate. The sulphate precipitate was converted to carbonate and dissolved in nitric acid. 234Th and 234Pa were eliminated by synergistic solvent extraction using tri-n-butyl phosphate and thenoyl trifluoroacetone extractant mixture in xylene. An iron scavenging step was included to remove any residual impurities. Finally, strontium is precipitated as SrC2O4 . H2O. The separated 90Sr activity was followed to check the equilibrium growth of 90Y.

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Abstract  

The identification of isotopic composition of depleted uranium obtained after the reprocessing of spent fuel is very much important as far as their further applications are concerned. The concentration of 235U in the reprocessed uranium will be lower and their depletion depends mainly on the type of the reactor and burn up. The depleted uranium has a variety of applications in nuclear fuel cycle operations and industries warranting quick assessment of its isotopic composition. This paper describes the methodology developed and used for the estimation of 235U concentration in reprocessed uranium in the form of uranyl nitrate solution. A non destructive technique by high resolution gamma-ray spectrometry with HPGe detector has been used for the analysis. The activity ratio of 238U/235U, obtained from the absolute activity measurement on the 185 keV gamma-line of 235U and 1001 keV gamma-line of 234mPa, has been used for the estimation of 235U isotopic content in the sample using a mathematical formula. This method offers rapid and reliable estimate of the 235U concentration in samples comparable with that of mass spectrometry measurements. The results show that 235U concentration in the sample can be determined within 5% error for 10000 seconds counting.

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Journal of Radioanalytical and Nuclear Chemistry
Authors: S. Ravi, A. Deepa, B. Surekha, S. Susheela, P. Achuthan, S. Anil Kumar, K. Vijayan, U. Jambunathan, S. Munshi, and P. Dey

Abstract  

90Sr estimation in reprocessed uranium was carried out by a series of solvent extraction and carrier precipitation techniques using strontium and lanthanum carriers. Fuming with HClO4 was used to remove 106Ru as RuO4. Three step solvent extraction with 50% tri-n-butyl phosphate in xylene in presence of small amounts of dibutyl phosphate and thenoyl trifluoro acetone was carried out to eliminate uranium, plutonium, thorium and protactinium impurities. Lanthanum oxalate precipitation in acid medium was employed to scavenge the remaining multivalent ions. Strontium was precipitated as strontium oxalate in alkaline pH and 137 Cs was removed by washing the precipitate with water. A strontium recovery well above 70% was obtained. Final estimation was carried out by radiometry using end window GM counter after drying the precipitate under an infra red lamp. The same procedure was extended to the estimation of 90Sr in a diluted sample of the actual spent fuel solution. An additional lanthanum oxalate precipitation step was required to remove the entire 144Ce impurity from this sample. This modified procedure was employed in the determination of 90Sr in a number of reprocessed uranium samples and the over all precision of the method was found to be well within ±10%. An additional barium chromate precipitation step was necessary for the analysis of reprocessed uranium samples from high bumup fuels to eliminate trace amounts of short lived 224Ra produced during the decay of 232U and its daughters as they interfere in the estimation of 90Sr.

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